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a,. es Caldwell - Fwd: FirstEnergy Nuclear Operating Company (FENOC) Exponent Failure Analysis Associates' Technidbio

From: John GrobeTo: Eric DuncanDate: Wed, May 9, 2007 10:09 AMSubject: Fwd: FirstEnergy Nuclear Operating Company (FENOC) Exponent Failure AnalysisAssociates' Technical Report

Eric,

You may want to send this to the exponent group

>>> "FERTEL, Marvin" <[email protected]> 05/09/2007 10:29 AM >>>May 9, 2007

I!

/

/

Mr. Luis A. Reyes

Executive Director for Operations

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

sy

Dear Mr. Reyes:

The FirstEnergy Nuclear Operating Company (FENOC) docketed a May 2,2007, letter responding to four specific issues raised by the NRC staffregarding the Exponent Failure Analysis Associates' technical report,"Review and Analysis of the Davis-Besse March 2002 Reactor PressureVessel Head Wastage Event." In the May 2nd letter, FENOC noted thatthey provided the Exponent report to NEI with a request that anevaluation be conducted to determine if the report calls into questionthe adequacy of the industry's operational monitoring or inspectionprograms or otherwise raise a potential generic safety concern.

Enclosure

Marvin S. Fertel

Senior Vice President and Chief Nuclear Officer

7Y~ /

James Caldwell - Fwd: FirstEnergy Nuclear Operating Company (FENOC) Exponent Failure Analysis Associates' Techni(•eP

Nuclear Energy Institute

1776 I Street NW, Suite 400

Washington, DC 20006

www.nei.org <http:llwww.nei.orq/>

P: 202-739-8125

F: 202-293-3451

E: msftcnei.orq

nuclear, clean air energy.

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Summary ReportMay 2007

Effect of Exponent Analysis of

Davis-Besse Reactor Vessel Head Wastage Event

On Industry Inspection Programs

Summary Report

Nuclear Energy Institute1776 1 Street, NW, Suite 400

Washington, D.C. 20006-3708


Contents

1. Panel Charter and Conclusions

2. Technical Support to the Position that Exponent Analysis does not Invalidate Industry RPVHead Inspection Program

3. References

-i-


1. Panel Charter and Conclusions

NEI commissioned an expert panel to review the First Energy Nuclear Operation Company(FENOC) report of December 15, 2006, authored by Exponent, entitled "Review and Analysis ofthe Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event" to determine thepotential impact on industry inspection programs for reactor vessel top head nozzles. The twomain questions addressed by the panel and conclusions are as follows:

I. Do the crack growth rates and RPV head wastage mechanisms identified in the Exponentreport call into question the adequacy of the industry's operational monitoring and periodicinspection requirements?

Panel Response: No, the crack growth rates and RPV head wastage mechanisms identifiedin the Exponent report do not call into question the industry's monitoring and inspectionrequirements. Specifically,

" The reported crack growth rates are near the upper end but within the industry datafor PWSCC of Alloy 600 material as documented in MRP-551, and the wastagemechanisms and rates are within the bounds defined by the EPRI Boric AcidCorrosion Guidebook2 in 1995 and Revision I to the guidebook in 200 13.

* Industry operational and monitoring programs (NRC EA-03-009 4 and ASME CodeCase N-729-1 5) are capable of preventing the type of condition postulated in theExponent report.

o Non-destructive examinations are specified at intervals appropriate to thehead temperature and service time to detect cracks in nozzle walls beforethey grow to leaks.

o Bare metal visual examinations of the vessel heads are specified at intervalsappropriate to the head temperature and service time to detect leaks at anearly stage.

o The combination of NDE and bare metal visual examinations providesprotection against large volumes of wastage from occurring.

o Boric acid corrosion programs meeting the requirements of NRC GenericLetter 88-056 have been implemented at all plants. These programs willprovide adequate advance warning of leakage and wastage.

o As a result of the Davis-Besse incident, the industry is far more sensitive tothe risk of boric acid corrosion, further decreasing the likelihood of Davis-Besse conditions not being detected and acted on in a timely manner.

1-I


2. Does the information in the Exponent report raise a (new) generic safey concern?

Response: No, the information in the Exponent report does not raise a new generic safetyconcern. Specifically,

" The postulated crack growth rates are near the upper end but within the distributionconsidered for nozzle cracking.

" The postulated head wastage rates are consistent with the upper bounds of boric acidcorrosion and subsequent wastage.

" The likelihood of the high stress intensity factors, high crack growth rates andmaximum corrosion rates, described in the Exponent report occurringsimultaneously is deemed small. Implementation of the inspection requirementscontained in the NRC Order EA-03-009, and/or ASME Code Case N-729-1,implementation of an effective boric acid corrosion inspection program per NRCGeneric Letter 88-05 and WCAP-1 59887, as well as timely reaction to plantindicators such as area radiation monitor filter clogging, RCS leakage detection, etc.,are expected to preclude a gross reactor vessel head wastage event such as thatwhich occurred at Davis-Besse.

Further supporting information to these conclusions are provided in the following section of thisreport.

- Section 2 provides technical support to the position that the industry periodicinspection programs are adequate to prevent the type of degradation observedat Davis-Besse given that the model and timeline proposed in the Exponentreport

1-2


2. Technical Support to Position that Exponent Analysis does not Invalidate IndustryRPV Head Inspection Programs

The purpose of this section is to provide technical support to the position that the Exponentreport "Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel HeadWastage Event" dated December 15, 2006 does not call into question the adequacy of theindustry's operational monitoring and periodic inspection requirements for reactor pressurevessel (RPV) top heads in pressurized water reactors (PWRs).

Summary of Exponent Proposed Leakage/Wastage ModelThe Exponent proposed model is essentially the same as described in the root cause analysissubmitted to the NRC on April 18, 2002 and August 27, 20028 with the following mainexceptions:

Leakage from Nozzle 3 occurred over a period of approximately one fuel cycle ratherthan the approximately 6-8 years in the original root cause report. The more rapidprogression of the leak results from the crack growth rate in the Nozzle 3 material beingfound by test to be up to four times higher than the 75h percentile crack growth rate forAlloy 600 material as described in MRP-55.1 The leak rate, and boric acid corrosion rate,increased suddenly about October/November 2001 when corrosion in the annulus froman axial crack in the CRDM nozzle uncovered a pre-existing crack through the J-grooveweld and increased again when the wastage uncovered the back side of the J-weld..

" Other indications of boric acid leakage (boric acid on the vessel head, containment aircooler cleaning, radiation monitor fouling, etc.) up through 12RFO (2000) were the resultof other leaks in containment, and not the result of the CRDM nozzle leaks which causedthe large corrosion volume. For example, accumulation of boric acid on the vessel headis attributed to leaking CRDM flange gaskets.

* The sequence of events that occurred in late 2001 leading to the large volume ofcorrosion was "unexpected and unpredictable."

Does Information in the Exponent Report Represent a New Generic Safety Concern?The panel tasked with evaluating the Exponent report has concluded that the report does notidentify a new generic safety concern. Specifically,

1. Reported Crack Growth RatesExponent states that the crack growth rates of the Davis-Besse CRDM nozzle material wereup to four times the growth rate of the MRP recommended 75th percentile crack growthcurve reported in MRP-55. While this statement is correct, Figure 2-1 from MRP-55shows that some of the material specimens in the MRP-55 database had crack growth rateseven higher than the Davis-Besse material. Therefore, the Davis-Bess material isconsidered to be within the bounds of previously evaluated materials.

2-1


2. Deterministic vs. Probabilistic AnalysesIndustry models for crack growth and wastage are based on probabilislic methods ratherthan the deterministic analyses performed by Exponent. Probabilistic analysis models andanalysis results are reported in MRP-110. 9 These analyses consider the full range of crackgrowth rates, leak rates and boric acid corrosion rates, including the higher growth rates.Therefore, this is not considered a new, unexpected, or unpredictable occurrence.Probabilistic models were used to establish inspection intervals.

3. Boric Acid Corrosion RatesData in the original EPRI Boric Acid Corrosion Guidebook2 in 1995 showed peakcorrosion/wastage rates of up to 10 inches/year for severe cases involving concentratedboric acid or impingement. Therefore, the more recent test data from Argonne, showingcorrosion rates of up to 7 inches/year, does not represent a new, unexpected, orunpredictable occurrence. The lower 2-4 inches/year estimated in the Davis-Besse rootcause evaluation report was considered a best estimate based on the sum total of availableevidence in the spring of 2002.

In summary, industry programs have already addressed the range of conditions hypothesized inthe Exponent report.

Current Inspection Requirements for Reactor Pressure Vessel (RPV) Top HeadsIn February 2003, the NRC issued EA-03-009 (subsequently revised by Order EA-03-009, Rev1, in Feb, 2004)4 establishing interim inspection requirements for RPV heads at pressurizedwater reactor plants. The industry has subsequently developed alternate requirements for RPVhead inspections that have been incorporated into the ASME Boiler and Pressure Vessel Codethrough Code Case N-729-1.5 These documents require both nondestructive examinations(NDE) for cracks and bare metal visual examinations (BMV) of the vessel head surface atintervals that were established based on the plant head temperature and operating time, whichhave been demonstrated statistically to correlate with the occurrence of RPV top head nozzlecracking. The primary intent of the nondestructive examinations is to ensure an acceptably smallprobability of leaks occurring in the nozzles, while the intent of the bare metal visualexaminations is to serve as a backup for the nondestructive examinations. There has also been asignificant increase in plants' sensitivity to the potential damage that can result from leakage andfar greater attention to changes in the plant unidentified leak rate.

The NRC order establishes three susceptibility categories (high, moderate and low) for RPVheads based on their operating time and temperature, characterized in terms of EffectiveDegradation Years (EDYs). EDYs correspond to the equivalent operating time of the head if itwere to operate at a reference temperature of 6007F. Corrections from the actual headtemperature to the reference temperature of 600°F are based on an Arrhenius-type relationshipderived from laboratory data. A statistical correlation of RPV top head cracking to EDYs isdocumented in MRP-105'0 . Plants in the high susceptibility category (> 12 EDYs) must performNDE and BMVs every refueling outage until they are replaced with heads fabricated withPWSCC-resistant materials. For plants in the moderate category (12 > EDYs > 8) BMV isrequired every outage and NDE every other outage. For plants in the low susceptibility category,BMVs are required every third refueling outage or 5 years, and NDE at least every 4 refueling

2-2


outages or 7 years. Over time, assuming heads are not replaced, the plants gradually progress tohigher categories, with more frequent inspections. As a result of inspection costs, and the needto plan for possible repairs, essentially all plants in the high susceptibility category and many inthe moderate category, have replaced or scheduled replacement of their original RPV heads withnew heads manufactured with PWSCC-resistant materials. For plants with replaced heads, theorder requires an inspection regimen similar to that for low susceptibility plants.

The industry-proposed inspection program, recently issued as Code Case N-729-15 , also uses atime-temperature Arrhenius correlation to establish Reinspection Years (RIYs). RIYs are similarto EDYs, but are calculated more conservatively, based on laboratory crack growth datacorrelations (rather than crack initiation). Except for the lowest susceptibility heads (see below),BMVs are required by the Code Case each refueling outage. NDE inspections are required bythe Code Case every 2.25 RIYs or eight calendar years, whichever is less. For heads with lessthan 8 EDYs, the Code Case permits BMVs to be extended to every third refueling outage or fiveyears, which is identical to the NRC order. The Code Case has been published by ASME, and isexpected to be accepted by the NRC (with some conditions) in the near future.

In its evaluation of the adequacy of current inspection programs, the panel assumed that all RPVheads in domestic PWR plants will be examined in the future based on either the NRC Order orthe ASME Code Case. As a result, the panel concludes that it is highly unlikely, given theabove-described inspection regimes and associated head replacements, that a plant wouldexperience the type of significant boric acid corrosion found at Davis-Besse in the future.

Evaluation of RPV Top Head Nozzle Examination Requirements Using Exponent ProposedCrack Growth RatesAs previously noted, the RPV top head safety evaluations, as documented in MRP- 1102, werebased on probabilistic predictions of crack growth, using data which encompass the Davis-Bessecrack growth rates referenced in the Exponent report.

As additional confirmation of the industry inspection program, deterministic evaluations wereperformed for specific plants using the crack growth curves proposed by Exponent for the 1/2TDavis-Besse specimens. These calculations were performed for the cases of a relatively hightemperature head, an intermediate temperature head, and a low temperature head. The threehead temperatures were chosen so as to conservatively bound the heads in each category thatremain in service in the domestic PWR fleet. The resulting predictions were then compared tothe NDE and BMV examinations required by NRC Order EA-03-009, Rev 1 and ASME CodeCase 729-1. Since the current industry requirements are aimed primarily towards discoveringcracks before they grow through-wall, the calculations are assumed to start at the approximatelimit of NDE detectability rather than the point where an initial leak occurs. A curve for thestress intensity factor versus crack length and a curve for crack growth rate representative of theDavis-Besse Nozzle 3 material were taken directly from the Exponent report.

Table 2-1 below provides a summary of the three cases analyzed, and the associated inspectionrequirements for each case, based on the NRC Order and the Code Case.

2-3


Table 2-1Description of Case Studies to Evaluate Industry Inspection Requirements

Case High Temperature Head Moderate Temperature Low Temperature HeadHead

Head Temp. 592 580 561(OF)

Current 14.5 25.3 10.8EFPYsCurrenteD t 10.5 11.1 2.1EDYs

Exams per Exams per Exams per Exams per Exams per Exams perOrder CC N-729-1 Order CC N-729-1 Order CC N-729-1

0* BMV+NDE BMV+NDE BMV+NDE BMV+NDE BMV+NDE BMV+NDE1 BMV BMV BMV BMV None None2 BMV+NDE BMV+NDE BMV+NDE BMV None None3 BMV+NDE BMV BMV+NDE BMV+NDE BMV BMV4 BMV+NDE BMV+NDE BMV+NDE BMV BMV+NDE None5 BMV+NDE BMV BMV+NDE BMV None BMV+NDE6 BMV+NDE BMV+NDE BMV+NDE BMV+NDE None None7 BMV+NDE BMV BMV+NDE BMV BMV None8 BMV+NDE BMV+NDE BMV+NDE BMV BMV+NDE BMV9 BMV+NDE BMV BMV+NDE BMV+NDE None None

* RFO 0 represents the time of the most recent NDE inspection at which it is conservatively assumed thata crack exists in the nozzle that is just smaller than the limit of NDE detection. This crack would bepredicted to reach detectable size immediately after the plant begins operation following thatinspection.

Deterministic predictions of crack growth for the three sample cases, using the crack growth ratecurve for Davis-Besse Nozzle 3 referenced in the Exponent report are presented in Figures 2-2,2-3 and 2-4. In each case, the crack growth rate equation was adjusted to the appropriate headoperating temperature in accordance with the Arrhenius equation for crack growth (MRP-55').The figures also show the predicted crack growth in accordance with the MRP-55 75 " percentilecrack growth law for comparison purposes. Several significant crack sizes are shown ashorizontal dashed lines on the figures, corresponding to the crack length at which detection byUT would be expected with high confidence (conservatively assumed = 0.25" ), the crack lengthcorresponding to the crack at the top of the weld, at which leakage is predicted, and a cracklength of 1.2" above the weld, corresponding to the crack length that was discovered at Davis-Besse in 2002. Finally, vertical arrows are plotted on the figures corresponding to the NDE andBMV inspections that would be required at various refueling outages, per Table 2-1 above.Since inspections required per the NRC Order are somewhat more conservative than thoserequired by Code Case N-729-1, only those required per the Code Case are shown on the figures.This produces the most conservative evaluation of the potential effect of the Exponent crackgrowth rate, since it corresponds to the least amount of inspection that would possibly beperformed in the future. Also for conservatism, the initial baseline inspection was assumed tooccur at a time when the postulated crack is just below the UT inspection threshold, and wouldthus be missed.

2-4


Figures 2-2 through 2-4 show that for each hypothetical case an NDE + BMV would beperformed shortly after the postulated cracks are projected to reach the top of the welds, suchthat they would be well above the detection threshold of the ultrasonic NDE, and evidence ofleakage would likely be detected by the BMV as well. These exams would be performed, andthe cracking detected, well before the crack length approached the size predicted to have causedthe severe wastage observed in the Davis-Besse head. These results, plus the conservativebounding nature of the deterministic evaluation, were a major factor in the panel's conclusionregarding adequacy of the current industry RPV top head examination program.

ConclusionsThe main conclusions are that the Exponent report has not identified a new generic issue and thatcurrent industry inspection programs provide adequate protection against rapidly growing cracksand high rates of boric acid wastage as postulated in the Exponent report.

2-5


Figure 2-1Davis-Besse Crack Growth Rates Relative to MRP-55 Data

A A A A

10 -~ ------ _---

lool

~ _ _ _ _ _ - _ _- _ _

-Ai

Et A A L A A

a; 1- - - A -- A_U~~A A A A A______ _______

AA

- -- M5P 75th %-tiA

.. ... DB Nozzle #3 Curvefit

0.01 1

0 10 20 30 40 50 60

Stress Intensity Factor, MPa mA.5

2-6


Figure 2-2Projected Crack Growth and Industry Examination Requirements for a High TemperatureHead under Crack Growth Rate Conditions Postulated in Exponent Report

2-7


Figure 2-3Projected Crack Growth and Industry Examination Requirements for 21 ModerateTemperature Head under Crack Growth Rate Conditions Postulated in Exponent Report

Evaluation of CCo729-1 Exam Regimenfor Moderate Temperature Head (580 F)

•E

cjrack 1.2" aboveAeer wn~tnneA

veld

.4.

2

,.1.5

0 1

0.5

0-

Crack and LeakDetected

K /

I______ Ii DB Nozz3 @ 580 F

-MRP-55 CGR (75%tle)

I ICrack at to(= Leakage

of weld

UT DetectioThreshold

ND

-F -DI

BBMV +NDE BMi~vI BMV

BMV-NDE

-I---

I0.0 1.5 3.0 4.5 6.0

Operating Tme (yrs.)

7.5 9.0 10.5 12.0

2-8


Figure 2-4Projected Crack Growth and Industry Examination Requirements for ai Low TemperatureHead under Crack Growth Rate Conditions Postulated in Exponent Report

Evaluation of CC-729-1 Exam Regimenfor Cold Head (561 F)

2.5 /Crak 1.2" above veld(sev re wastage)

2 ______

Crack and Leak F- @Detected DNo3@56F

1.5 ---... i--RP-55 CGR (75%file)

C rack at top of weld(:Leakage)

0.0 1.5 3.0 4.5 6.0 7.5 , 9.0 10.5 12.0 13.5 15.0

Operating Tine (yrs.)

2-9

I .


3. References

1. Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary WaterStress Corrosion Cracking (PWSCC) of Thick- Wall Alloy 600 Materials (MRP-55) Revision1, EPRI, Palo Alto, CA: 2002, 1006695.

2. Boric Acid Corrosion Guidebook: Recommended Guidance for Addressing Boric AcidCorrosion and Leakage Reduction Issues, EPRI, Palo Alto, CA: 1995, TR-102748.

3. Boric Acid Corrosion Guidebook, Revision 1: Managing Boric Acid Corrosion Issues atPWR Stations, EPRI, Palo Alto, CA: 2001, 1000975

4. Issuance of First Revised NRC Order (EA-03-009) Establishing Interim InspectionRequirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors, Feb. 20,2004.

5. Code Case N-729, Rev. 1, Alternative Examination Requirements for PWR Closure Heads

with Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section A7, Division I.

6. Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWRPlants, NRC Generic Letter 88-05, 3/17/1988.

7. Generic Guidance for an Effective Boric Acid Inspection Progran for Pressurized WaterReactors, Westinghouse, March 2003, WCAP-15988.

8. Root Cause Analysis Report: Significant Degradation of the Reactor Vessel Head - Revision1, First Energy, Davis-Besse Nuclear Power Station, 8/27/2002, CR 2002-0891.

9. Materials Reliability Program Reactor Vessel Closure Head Penetration Safety Assessmentfor U.S. PWR Plants (MRP-110): Evaluations Supporting the MRP Inspection Plan, EPRI,Palo Alto, CA: 2004, 1009807.

10. Materials Reliability Program Probabilistic Fracture Mechanics Analysis ofPWR ReactorPressure Vessel Top Head Nozzle Cracking (MRP-105), May, 2004

11. J.F. Hall, Boric Acid Corrosion ofLow Alloy Steel, presented at EPRI Workshop on PrimaryWater Stress Corrosion Cracking ofAlloy 600 in PWRs, Charlotte, NC, October 9-11, 1991.

3-1

NUCLEAR ENERGY INSTITUTE

Marvin S. FertelSENIOR VICE PRESIDENT ANDCHIEF NUCLEAR OFFICER

May 9, 2007

Mr. Luis A. ReyesExecutive Director for OperationsU.S. Nuclear Regulatory CommissionWashington, DC 20555-0001

Dear Mr. Reyes:

The FirstEnergy Nuclear Operating Company (FENOC) docketed a May 2, 2007, letter responding tofour specific issues raised by the NRC staff regarding the Exponent Failure Analysis Associates'technical report, "Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel HeadWastage Event." In the May 2nd letter, FENOC noted that they provided the Exponent report to NEIwith a request that an evaluation be conducted to determine if the report calls into question theadequacy of the industry's operational monitoring or inspection programs or otherwise raise apotential generic safety concern.

NEI commissioned an expert panel to conduct an evaluation of the report. The findings andconclusions were reviewed with the NEI Materials Executive Oversight Group (MEOG) responsible foroversight and coordination of the industry programs involving management of materials issues. TheMEOG concurred with the expert panel's findings. NEI also briefed the industry Chief NuclearOfficers of the evaluation and its conclusions. NEI provided the following in response to FENOC'srequest:

1. Do the crack growth rates and reactor pressure vessel (RPV) head wastage mechanismsidentified in the report call into question the adequacy of the industry's monitoring andinspection programs?

Response: No. We believe the industry's materials monitoring programs are sound and willhelp maintain safe operation of nuclear power plants. The reported crack growth rates arewithin the industry data for primary water stress corrosion cracking of Alloy 600 materialsdocumented In the EPRI Materials Reliability Program technical report, "Materials ReliabilityProgram Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking(PWSCC) of Thick-Wall Alloy 600 Materials (MRP-55)," Revision 1., TR 1006695. Further, thewastage mechanisms and rates are within the bounds described in the EPRI Boric AcidCorrosion Guidebook (TR - 102748).

Mr. Luis A. ReyesMay 9, 2007Page 2

The industry's operational monitoring and inspection programs as they relate to managingdegradation of Alloy 600 nozzles located in the RPV head are capable of preventing the typeof conditions postulated the Exponent report.

2. Does the information in the report raise a potential generic Safety concern?

Response: No. The expert panel reviewed the reported crack growth rates and RPV headwastage analyses and concluded there is no new potential generic safety concern. Plantsafety is not jeopardized because the postulated crack growth rates are within thedistribution considered for nozzle cracking and the wastage rates are consistent with upperbounds of boric acid corrosion. This coupled with the industry's operational monitoring andinspection programs will continue to assure plant safety.

I have enclosed for your information a copy of the expert panel's summary report containing furtherdetails supporting the responses summarized above. Please contact me directly or Jay K. Thayer at202.739.8112, ikt5nei.orc should you have any questions.

Sincerely,

Marvin S. Fertel

Enclosure

c: William F. Kane, Deputy Executive Director, NRCJames E. Dyer, Director, NRCJohn A. Grobe, Associate Director, NRCPublic Document Room

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